U.S. patent application number 14/451703 was filed with the patent office on 2015-02-05 for integral molten salt reactor.
The applicant listed for this patent is TERRESTRIAL ENERGY INC.. Invention is credited to David LEBLANC.
Application Number | 20150036779 14/451703 |
Document ID | / |
Family ID | 52427662 |
Filed Date | 2015-02-05 |
United States Patent
Application |
20150036779 |
Kind Code |
A1 |
LEBLANC; David |
February 5, 2015 |
INTEGRAL MOLTEN SALT REACTOR
Abstract
The present relates to the integration of the primary functional
elements of graphite moderator and reactor vessel and/or primary
heat exchangers and/or control rods into an integral molten salt
nuclear reactor (IMSR). Once the design life of the IMSR is
reached, for example, in the range of 3 to 10 years, it is
disconnected, removed and replaced as a unit. The spent IMSR
functions as the medium or long term storage of the radioactive
graphite and/or heat exchangers and/or control rods and/or fuel
salt contained in the vessel of the IMSR. The present also relates
to a nuclear reactor that has a buffer salt surrounding the nuclear
vessel. During normal operation of the nuclear reactor, the nuclear
reactor operates at a temperature that is lower than the melting
point of the buffer salt and the buffer salt acts as a thermal
insulator. Upon loss of external cooling, the temperature of the
nuclear reactor increases and melts the buffer salt, which can then
transfer heat from the nuclear core to a cooled containment
vessel.
Inventors: |
LEBLANC; David; (Ottawa,
CA) |
|
Applicant: |
Name |
City |
State |
Country |
Type |
TERRESTRIAL ENERGY INC. |
Mississauga |
|
CA |
|
|
Family ID: |
52427662 |
Appl. No.: |
14/451703 |
Filed: |
August 5, 2014 |
Related U.S. Patent Documents
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Application
Number |
Filing Date |
Patent Number |
|
|
61862378 |
Aug 5, 2013 |
|
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Current U.S.
Class: |
376/207 |
Current CPC
Class: |
G21C 5/02 20130101; Y02E
30/00 20130101; G21D 1/00 20130101; G21C 19/28 20130101; G21C
17/112 20130101; G21C 1/322 20130101; G21D 3/00 20130101; G21D 9/00
20130101; Y02E 30/30 20130101; G21C 1/22 20130101 |
Class at
Publication: |
376/207 |
International
Class: |
G21C 7/00 20060101
G21C007/00 |
Claims
1. A method of operating a nuclear power plant, the nuclear power
plant comprising a nuclear reactor to produce heat, a heat
exchanger system, and an end use system, the heat exchanger system
to receive heat produced by the nuclear reactor and to provide the
received heat to the end use system, the method comprising steps
of: operating the nuclear reactor, the nuclear reactor comprising a
vessel and a graphite moderator core positioned in the vessel, the
heat exchanger system having an inside portion located inside the
vessel and an outside portion located outside the vessel; shutting
down the nuclear reactor upon occurrence of a shutdown event, to
obtain a shutdown nuclear reactor; severing all operational
connections between the inside portion of the heat exchanger system
and the outside portion of the heat exchanger system to obtain a
severed, shutdown nuclear reactor; obtaining a replacement nuclear
reactor having an inner heat exchanger system portion; and
operationally connecting a replacement nuclear reactor to the
outside portion of heat exchanger system by connecting the inner
heat exchanger system portion of the replacement nuclear reactor to
the outside portion of the heat exchanger system.
2. The method of claim 1 wherein the shutdown event includes at
least one of: strain in the graphite core exceeding a threshold
strain; a calculated fast neutron fluence on the graphite core
exceeding a threshold fast neutron fluence; and an operation
duration of the nuclear reactor exceeding a pre-determined
operation duration.
3. The method of claim 1 further comprising the step of
sequestering the severed, shutdown nuclear reactor.
4. The method of claim 3 wherein sequestering the severed, shutdown
nuclear reactor is preceded by a step of leaving the severed,
shut-down nuclear reactor in place to allow the severed, shutdown
nuclear reactor to cool by radioactive decay of radioactive
elements present in the severed, shut-down nuclear reactor.
5. The method of claim 2 wherein the graphite moderator core has a
damage operation duration beyond which the graphite moderator core
becomes damaged, the pre-determined duration of operation being
shorter than the core damage operation duration.
6. The method of claim 1 wherein: the nuclear reactor is a molten
salt reactor (MSR) that runs on a molten fuel salt; the nuclear
power plant further comprises radioactivity detectors, and shut-off
mechanisms, the graphite moderator core defines one or more than
one through hole, and the inside portion of the heat exchanger
system comprises: a heat exchanger unit, the heat exchanger unit
having a plurality of heat exchangers arranged therein, each heat
exchanger having a coolant salt circulating therein, the heat
exchanger unit being in fluid communication with the one or more
than one through hole of the graphite moderator core, the MSR
further comprises: a pump system to pump the molten fuel salt
through the heat exchanger unit and through the one or more than
one through hole of the graphite moderator core, the heat
exchangers being arranged in the heat exchanger unit to have the
molten fuel flow thereon, each heat exchanger having associated
thereto a respective radioactivity detector, each radioactivity
detector arranged to detect radioactivity present in the coolant
salt circulating in the respective heat exchanger, and each heat
exchanger having associated thereto a respective shutoff mechanism
arranged to shut off circulation of the coolant salt circulating in
the respective heat exchanger, the method further comprising, prior
to shutting down the nuclear reactor, activating the shutoff
mechanism of a particular heat exchanger when radioactivity beyond
a threshold amount is detected, by the radioactivity detector of
the particular heat exchanger, in the particular heat
exchanger.
7. The method of claim 6 wherein: each respective heat exchanger
has an inlet conduit and an outlet conduit, and severing any
operational connection between the inside portion of the heat
exchanger system and the outside portion of the heat exchanger
system includes severing the inlet conduit and the outlet conduit
of each heat exchanger.
8. The method of claim 6 wherein: the shutoff mechanism includes a
crimping mechanism to crimp the particular heat exchanger to
prevent the coolant salt from flowing therethrough; and activating
the shutoff mechanism of the particular heat exchanger includes
crimping the particular heat exchanger.
9. The method of claim 6 wherein: the shutoff mechanism includes a
valve mechanism to close off coolant salt flow in the particular
heat exchanger; and activating the shutoff mechanism of the
particular heat exchanger includes closing the valve of the
particular heat exchanger.
10. The method of claim 6 wherein: the shutoff mechanism includes a
freezing mechanism to freeze a portion of the particular heat
exchanger to prevent the coolant salt from flowing therethrough;
and activating the shutoff mechanism of the particular heat
exchanger includes freezing the portion of the particular heat
exchanger.
11. The method of claim 1 wherein: the nuclear reactor is a molten
salt reactor (MSR) that runs on a molten fuel salt; the nuclear
power plant further comprises radioactivity detectors, and shut-off
mechanisms, the graphite moderator core defines one or more than
one through hole, and the inside portion of the heat exchanger
system comprises: a heat exchanger unit, the heat exchanger unit
having a plurality of heat exchangers arranged therein, each heat
exchanger having a coolant salt circulating therein, the heat
exchanger unit being in fluid communication with the one or more
than one through hole of the graphite moderator core, the MSR
further comprises: a pump system to pump the molten fuel salt
through the heat exchanger unit and through the one or more than
one through hole of the graphite moderator core, the heat
exchangers being arranged in the heat exchanger unit to have the
molten fuel flow thereon, and a pressure monitoring system, each
heat exchanger being operationally connected to the pressure
monitoring system, the pressure monitoring system to monitor the
pressure of the coolant salt circulating in the respective heat
exchanger, each heat exchanger having associated thereto a
respective shutoff mechanism arranged to shut off circulation of
the coolant salt circulating in the respective heat exchanger, the
method further comprising, prior to shutting down the MSR,
activating the shutoff mechanism of the particular heat exchanger
when the pressure monitoring system detects a pressure change in
the particular heat exchanger.
12. The method of claim 1 wherein the nuclear reactor is a molten
salt reactor (MSR) that runs on a molten fuel salt disposed
therein, the nuclear power plant further comprising a dump tank
operationally connected to the vessel, the method further
comprising: subsequent to severing any operational connection
between the inside portion of the heat exchanger system and the
outside portion of the heat exchanger system, emptying the molten
fuel salt from the nuclear reactor into the dump tank.
13. The method of claim 12 further comprising transferring the
molten salt from the dump tank to the replacement nuclear
reactor.
14. A nuclear reactor unit comprising: a containment vessel; a
nuclear reactor located in the containment vessel, the nuclear
reactor having a reactor vessel that has a reactor vessel wall; and
a buffer salt contained in the containment vessel, the buffer salt
being in thermal contact with the reactor vessel wall, the nuclear
reactor, when running, to generate a heat output that produces a
first reactor vessel wall temperature, the buffer salt being in a
solid state when at a temperature equal to or below the first
reactor vessel wall temperature, the nuclear reactor, when external
cooling is lost, to generate heat that produces a second reactor
vessel wall temperature greater than the first reactor vessel wall
temperature, the buffer salt to absorb a portion of the decay heat,
an absorption of the portion of the decay heat to raise the
temperature of the buffer salt, the buffer salt to melt and become
a liquid buffer salt when at the second reactor wall temperature,
the containment vessel to maintain the liquid salt in thermal
contact with the reactor vessel wall.
15. The nuclear reactor of claim 14 wherein convective heat
transfer in the liquid state is higher than conductive heat
transfer in the solid state.
16. The nuclear reactor of claim 14 wherein the buffer salt is a
thermal insulator in the solid state and a thermal conductor in the
liquid state.
17. The nuclear reactor of claim 14 wherein the liquid buffer salt
conducts heat between the reactor vessel and the containment
vessel.
18. The nuclear reactor of claim 17 wherein the containment vessel
is in thermal contact with an exterior heat absorbing material.
19. The nuclear reactor of claim 18 wherein the exterior heat
absorbing material includes water.
20. The nuclear reactor of claim 14 wherein the containment vessel
comprises an inner wall and an outer wall, the inner wall being in
thermal contact with the reactor vessel wall, the buffer salt being
located between the inner wall and the outer wall.
21. The nuclear reactor of claim 14 wherein the nuclear reactor is
a molten salt nuclear reactor.
22. A nuclear power plant comprising: a molten salt reactor (MSR)
to produce heat, the MSR reactor comprising a vessel and a graphite
moderator core positioned in the vessel; a heat exchanger system
having a coolant salt circulating therein; a strain sensor arranged
to measure strain in the graphite moderator core; and an end use
system, the heat exchanger system to receive heat produced by the
nuclear reactor and to provide the received heat to the end use
system, the strain sensor to provide a signal indicative of
excessive strain when the strain in the graphite moderator core
exceeds a strain threshold value.
23. The nuclear power plant of claim 22 wherein: the graphite
moderator core defines one or more than one through hole, the heat
exchanger system further comprises heat exchangers disposed in the
vessel, about a longitudinal axis of the vessel, each heat
exchanger having a coolant salt circulating therein, the heat
exchangers being in fluid communication with the one or more than
one through hole of the graphite moderator core, the heat
exchangers defining an opening above the graphite moderator core,
and the vessel comprises a baffle structure positioned between the
heat exchangers and the graphite moderator core, the baffle
structure to guide molten salt fluid flowing downwards in the
vessel and out of the heat exchangers, along an outside periphery
of the graphite moderator core;
24. A nuclear power plant comprising: a molten salt reactor (MSR)
to produce heat; a heat exchanger system; radioactivity detectors
positioned outside the vessel; shutoff mechanisms positioned
outside the vessel; and an end-use system, the MSR comprising a
vessel, a graphite moderator core positioned in the vessel, and a
molten salt circulating at least in the vessel, the molten salt to
transfer the heat produced by the MSR to the heat exchanger system,
the graphite moderator core defining one or more than one through
hole, the heat exchanger system to receive the heat produced by the
MSR and to provide the received heat to the end use system, the
heat exchanger system comprising a plurality of heat exchangers in
fluid communication with the one or more than one through hole of
the graphite moderator core, each heat exchanger having associated
thereto a respective radioactivity detector, each radioactivity
detector arranged to detect radioactivity present in the coolant
salt circulating in the respective heat exchanger, each shutoff
mechanism arranged to shut off circulation of the coolant salt
circulating in the respective heat exchanger when radioactivity
beyond a threshold amount is detected, by the respective
radioactivity detector, in the respective heat exchanger.
25. The nuclear power plant of claim 24 further comprising a pump
system to pump the molten fuel salt through the heat exchanger
system and through the one or more than one through hole of the
graphite moderator core, the heat exchangers being arranged in the
heat exchanger unit to have the molten fuel flow thereon;
26. The nuclear reactor plant of claim 24 further comprising a
pressure monitoring system positioned outside the vessel, the
pressure monitoring system arranged to monitor a pressure of a
coolant salt circulating in each of the heat exchangers, each
shutoff mechanism arranged to shut off circulation of the coolant
salt circulating in the respective heat exchanger when the pressure
monitoring system detects a pressure change in the respective heat
exchanger.
27. The nuclear power plant of claim 24 wherein each respective
heat exchanger has an inlet conduit and an outlet conduit, each
respective radioactivity detector of each respective heat exchanger
being arranged to detect radioactivity present in the coolant salt
passing through at least one of the inlet conduit and the outlet
conduit.
28. The nuclear power plant of claim 27 wherein the shutoff
mechanism of each respective heat exchanger is arranged to shut off
the flow of the coolant salt in each of the inlet conduit and the
outlet conduit.
29. The nuclear power plant of claim 28 wherein each shutoff
mechanism includes a crimping system to crimp the inlet conduit and
to crimp the outlet conduit.
30. The nuclear power plant of claim 28 wherein each shutoff
mechanism includes an inlet conduit shutoff valve arranged on the
inlet conduit and an outlet conduit shutoff valve arranged on the
outlet conduit.
31. The nuclear power plant of claim 24 wherein at least one heat
exchanger is positioned in the vessel, the vessel further
comprising a neutron reflector positioned therein, between the at
least one heat exchanger and the graphite neutron moderator core,
the neutron reflector defining at least one aperture fluidly
connecting the heat exchanger unit and the graphite neutron
moderator core.
32. The nuclear power plant of claim 24 wherein the shutoff
mechanisms further comprise sever mechanisms to physically sever
each inlet conduit and outlet conduit to allow sequestration of the
vessel and any segments of inlet conduits and outlet conduits
attached to the vessel.
Description
CROSS-REFERENCE TO RELATED APPLICATIONS
[0001] This application is a continuation-in-part of International
Patent Application No. PCT/CA2013/050090 filed Feb. 6, 2013, which
claims the benefit of U.S. Provisional Application No. 61/633,071,
filed Feb. 6, 2012. The present continuation-in-part claims the
benefit of U.S. Provisional Application 61/862,378, filed on Aug.
5, 2013. The contents of International Patent Application No.
CA2013/050090, of U.S. Provisional Application No. 61/633,071, and
of U.S. Provisional 61/862,378 are incorporated herein by
reference.
FIELD
[0002] The present disclosure relates generally to nuclear
reactors. More particularly, the present disclosure relates to
molten salt nuclear reactors.
BACKGROUND
[0003] Molten salt reactors (MSRs) were primarily developed from
the 1950s to 1970s but, as of late, there has been increasing world
interest in this type of reactor. Older concepts are being
re-evaluated and new ideas put forth. This class of nuclear reactor
has a great deal of advantages over current nuclear reactors, the
advantages including potentially lower capital costs, overall
safety, long lived waste profile and resource sustainability.
[0004] With MSRs advantages also come some significant
technological challenges which lead to difficult basic design
decisions. The first and likely foremost is whether and how a
neutron moderator may be employed. Graphite has, in almost all
cases, been chosen as a moderator as it behaves very well in
contact with the fluoride salts used in MSRs. These salts are
eutectic mixtures of fissile and fertile fluorides (UF4, ThF4, PuF3
etc) with other carrier salts such as LiF, BeF2 or NaF. Using
graphite as a bulk moderator within the core of the MSR has many
advantages. For example, it gives a softer or more thermalized
neutron spectrum which provides improved reactor control and a
greatly lowered starting fissile inventory. As well, using graphite
throughout the core of a MSR allows the ability to employ what is
known as an under-moderated outer zone which acts as a net absorber
of neutrons and helps shield the outer reactor vessel wall from
damaging neutron exposure. The vessel, which contains the nuclear
core, has typically been proposed as being made of a high nickel
alloy such as Hastelloy.RTM. N; however, other materials are
possible.
[0005] The use of graphite within the core of the MSR (i.e., within
the neutron flux of a MSR) can have a serious drawback however.
That is, that graphite will first shrink and then expand beyond its
original volume as it is exposed to a fast neutron flux. Overall
expansion of graphite (graphite core) occurs when the volume of the
graphite (graphite core) is larger than its original volume, i.e.,
the volume preceding any neutron irradiation. An upper limit of
total fast neutron fluence can be calculated and operation of the
MSR is such that this limit is not exceeded. This limit determines
when the graphite would begin to expand beyond its original volume
and potentially damage surrounding graphite elements or the reactor
vessel itself. How long graphite can be used within the reactor
core is thus directly related to the local power density and thus
to the fast neutron flux it experiences. A low power density core
may be able to use the same graphite for several decades. This is
the case for many previous reactors employing graphite such as the
British gas cooled Magnox and AGR reactors. They were extremely
large and had a low power density for thermohydraulic reasons but,
this permitted an extremely long graphite lifetime. However, MSRs
would benefit from having a far higher power density and thus
graphite lifetime can become an issue.
[0006] The scientists and engineers designing MSRs have long been
faced with important design options. A first option is to simply
design the reactor to be quite large and very low power density in
order to get a full 30 year or more lifetime out of the graphite.
Thus one can seal all the graphite within the vessel and the
graphite can remain in the vessel for the design life of the
nuclear plant. Examples of this choice can be found in the studies
of Oak Ridge National Laboratories (ORNL) in the late 1970s and
early 1980s. For example, ORNL.TM. 7207 proposes a 1000 MWe reactor
which was termed the "30 Year Once Through" design which would have
a large reactor vessel of approximately 10 meters in diameter and
height in order to avoid the need for graphite replacement. Much of
the later work by Dr. Kazuo Furukawa of Japan, on what are known as
the FUJI series of reactor designs, also chose this route of large,
low power nuclear cores. These very large cores have obvious
economic disadvantages in terms of the sheer amount of material
required to fabricate the core and reactor vessel, and in the
excessive weight of the core. These challenges increase the cost
and complexity of the surrounding reactor building as would be
understood by those trained in the field. It should be added that a
30 year nuclear plant lifetime was quite acceptable in the 1970s
but by today's standards would be thought short. 50 or 60 years is
now desired and would mean a still larger core to allow this
lifetime without graphite replacement.
[0007] A second option often proposed is to employ a much smaller,
higher power density core but to plan for periodic replacement of
the graphite. This approach was commonly assumed in the work at Oak
Ridge National Laboratories (ORNL) in the design of the Molten Salt
Breeder Reactor from about 1968 to 1976 before the program was
cancelled. This 1000 MWe reactor design had an outer vessel of
Hastelloy.RTM. N that would contain hundreds of graphite elements
fitting together and filling the vessel but with passage channels
for the molten salt fuel to flow and exit the core to external heat
exchangers. In this second option, the reactor has much smaller
dimensions which are of approximately 6 meters in diameter and
height. In this case the graphite, particularly in the center of
the core with the highest fast neutron flux, only had an expected
lifetime of 4 years. Thus the reactor had to be designed to be shut
down and opened up every 4 years to replace a large fraction of the
graphite elements. This may not sound overly difficult to those not
trained in the field but with molten salts, the fission products,
some of which are relatively volatile, are in the fuel salt and can
also embed themselves onto a surface layer of graphite and, for
example, the inner metal surfaces of the reactor vessel. Thus just
opening the reactor vessel was known to be an operation that could
be difficult to perform without allowing radioactive elements to
spread into the surrounding containment zone. As well, the design
of the reactor vessel itself is more complex when it needs to be
periodically opened. These challenges are why the route of larger,
lower power density cores were often chosen.
[0008] A third option is to try to omit the use of graphite
altogether. This is possible and results in reactors typically with
a much harder neutron spectrum. An example of this choice is the
Molten Salt Fast Reactor (MSFR) proposed by a consortium of French
and other European researchers starting around year 2005. It has
very serious drawbacks however. For example it requires upwards of
five times the starting fissile load and any accidental exposure of
the salt to a moderator, such as water or even hydrogen content in
concrete, could lead to criticality dangers.
[0009] Beyond the issue of graphite lifetime, there are also the
somewhat related issues of the lifetime of the reactor vessel
itself and of the primary heat exchangers.
[0010] The reactor vessel wall may also have a limited lifetime due
to neutron fluence with both thermal and fast neutrons potentially
causing problems. The most commonly proposed material being a high
nickel alloy, such as Hastelloy.RTM. N, with reasonably well
understood behaviour and allowed limits of neutron fluence. As
such, a great deal of effort goes into core design to limit the
exposure of neutrons and/or lower the operating temperature of the
vessel wall. As well, adding thickness to the wall may help as
strength is lost with increased neutron exposure. This adds both
weight and expense. It is thus a challenge to have a 30 to 60 year
lifetime of the reactor vessel itself.
[0011] Another design challenge is the primary heat exchangers
which transfer heat from the radioactive primary fuel salt to a
secondary coolant salt. This coolant salt then typically transfers
heat to a working media such as steam, helium, CO2 etc. In some
cases these heat exchangers are outside or external the reactor
vessel itself, which appears to be the case for all 1950s to 1980s
ORNL designs. They also may be located within the reactor vessel
itself which has its own set of advantages and challenges. One
great advantage of internal heat exchangers is no radiation of
significance need leave the reactor itself as only secondary
coolant salt enters and leaves the vessel.
[0012] For both internal and external heat exchangers, the great
challenge is in either servicing or replacing them. When a MSR is
opened up, it can potentially lead to radioactivity being released
into a containment zone or space. ORNL for example proposed common
tube in shell heat exchangers external to the core, four heat
exchanger units per 1000 MWe reactor. In the case of any tube leaks
the operation was not to fix or plug tubes but to open the shell
and remove the entire tube bundle and replace with a new bundle.
Only after a cooling period would a decision be made on repair and
reuse of the bundle or simple disposal. Thus it is clear that
primary heat exchanger service and/or replacement techniques are a
great challenge in MSR design.
[0013] Further, when either graphite or heat exchangers are
replaced, then the issue of their safe storage must be also
addressed as they will become significantly radioactive during
operation. This represents yet another challenge in MSR overall
plant design.
[0014] It should be further highlighted that the related nuclear
design field of Fluoride salt cooled, High temperature Reactors
(known as FHRs) has very similar issues. In this work the reactor
design can be very similar but instead of the fuel being in the
fluoride salt, it is in solid form within the graphite moderator
using the fuel form known as TRISO. In this case the limited
graphite lifetime is also a function of the lifetime of the solid
TRISO fuels; however, all other design issues and challenges are
very similar to MSR design work. In FHRs, the primary coolant salt
is not nearly as radioactive but does typically contain some
radioactive elements such as tritium and a similar set of
challenges are present when planning to use solid block TRISO fuels
and periodically replace them. A subset of FHR design involves
using a pebble fuel form which does ease fuel replacement without
opening up the reactor vessel; however, this type of design has its
own set of issues
[0015] The decay heat that follows the shutdown of a nuclear
reactor following the loss of external cooling has been a
long-standing industry challenge. The incident at Fukushima Japan
indicates the seriousness of the issue. If the decay heat is not
removed quickly from the reactor, the temperature in the reactor
rises to unacceptable levels. Thus the speed with which the initial
decay heat can be removed from the reactor is critical.
[0016] Therefore, improvements in nuclear reactors are
desirable.
SUMMARY
[0017] The present disclosure relates to the integration of the
primary functional elements of graphite moderator and reactor
vessel and/or primary heat exchangers and/or control rods into a
single replaceable unit having a higher and more economic power
density while retaining the advantages of a sealed unit. Once the
design life of such an Integral Molten Salt Reactor (IMSR) is
reached, for example, in the range of 3 to 10 years it is
disconnected, removed and replaced as a unit and this unit itself
may also potentially function as the medium or long term storage of
the radioactive graphite and/or heat exchangers and/or control rods
and/or fuel salt itself. The functions of decay heat removal and
volatile off gas storage may also be integrated in situ.
[0018] The present disclosure also relates to nuclear reactor that
has a reactor vessel surrounded by a buffer material. The buffer
material can absorb decay heat when external cooling is lost. The
absorption of decay heat is effected by the buffer material phase
transition latent heat, the phase transition being that of solid
phase to liquid phase. The absorption is also effected by
convective heat transfer when the buffer material is in the liquid
state. The convective heat transfer occurs between the reactor
vessel and a heat sink in thermal contact with the buffer
material.
[0019] In a first aspect of the disclosure, there is provided a
method of operating a nuclear power plant, the nuclear power plant
comprising a nuclear reactor to produce heat, a heat exchanger
system, and an end use system, the heat exchanger system to receive
heat produced by the nuclear reactor and to provide the received
heat to the end use system. The method comprising steps of:
operating the nuclear reactor, the nuclear reactor comprising a
vessel and a graphite moderator core positioned in the vessel, the
heat exchanger system having an inside portion located inside the
vessel and an outside portion located outside the vessel; shutting
down the nuclear reactor upon occurrence of a shutdown event, to
obtain a shutdown nuclear reactor; severing all operational
connections between the inside portion of the heat exchanger system
and the outside portion of the heat exchanger system to obtain a
severed, shutdown nuclear reactor; obtaining a replacement nuclear
reactor having an inner heat exchanger system portion; and
operationally connecting a replacement nuclear reactor to the
outside portion of heat exchanger system by connecting the inner
heat exchanger system portion of the replacement nuclear reactor to
the outside portion of the heat exchanger system.
[0020] In a second aspect of the disclosure, there is provided a
nuclear reactor unit that comprises: a containment vessel; a
nuclear reactor located in the containment vessel, the nuclear
reactor having a reactor vessel that has a reactor vessel wall; and
a buffer salt contained in the containment vessel. The buffer salt
is in thermal contact with the reactor vessel wall. The nuclear
reactor, when running, is to generate a heat output that produces a
first reactor vessel wall temperature. The buffer salt is in a
solid state when at a temperature equal to or below the first
reactor vessel wall temperature. The nuclear reactor, when
shutdown, is to generate decay heat that produces a second reactor
vessel wall temperature greater than the first reactor vessel wall
temperature. The buffer salt is to absorb a portion of the decay
heat, an absorption of the portion of the decay heat to raise the
temperature of the buffer salt, the buffer salt is to melt and
become a liquid buffer salt when at the second reactor wall
temperature. The containment vessel to maintain the liquid salt in
thermal contact with the reactor vessel wall.
[0021] In a third aspect, the present disclosure provides a nuclear
power plant that comprises: a molten salt reactor (MSR) to produce
heat, the MSR reactor comprising a vessel and a graphite moderator
core positioned in the vessel; a heat exchanger system having a
coolant salt circulating therein; a strain sensor arranged to
measure strain in the graphite moderator core; and an end use
system, the heat exchanger system to receive heat produced by the
nuclear reactor and to provide the received heat to the end use
system, the strain sensor to provide a signal indicative of
excessive strain when the strain in the graphite moderator core
exceeds a strain threshold value.
[0022] In a fourth aspect, the present disclosure provides a
nuclear power plant that comprises: a molten salt reactor (MSR) to
produce heat; a heat exchanger system; radioactivity detectors
positioned outside the vessel; shutoff mechanisms positioned
outside the vessel; and an end-use system, the MSR comprising a
vessel, a graphite moderator core positioned in the vessel, and a
molten salt circulating at least in the vessel, the molten salt to
transfer the heat produced by the MSR to the heat exchanger system,
the graphite moderator core defining one or more than one through
hole, the heat exchanger system to receive the heat produced by the
MSR and to provide the received heat to the end use system, the
heat exchanger system comprising a plurality of heat exchangers in
fluid communication with the one or more than one through hole of
the graphite moderator core, each heat exchanger having associated
thereto a respective radioactivity detector, each radioactivity
detector arranged to detect radioactivity present in the coolant
salt circulating in the respective heat exchanger, each shutoff
mechanism arranged to shut off circulation of the coolant salt
circulating in the respective heat exchanger when radioactivity
beyond a threshold amount is detected, by the respective
radioactivity detector, in the respective heat exchanger.
[0023] Other aspects and features of the present disclosure will
become apparent to those ordinarily skilled in the art upon review
of the following description of specific embodiments in conjunction
with the accompanying figures.
BRIEF DESCRIPTION OF THE DRAWINGS
[0024] Embodiments of the present disclosure will now be described,
by way of example only, with reference to the attached figures
[0025] FIG. 1A shows an embodiment of a molten salt nuclear reactor
in accordance with the present disclosure.
[0026] FIG. 1B shows a molten salt nuclear reactor operationally
connected to a dump tank.
[0027] FIG. 2 shows a top view of the embodiment of FIG. 1.
[0028] FIG. 3 shows, in accordance with the present disclosure,
inlet and outlet molten salt conduits arranged to be shutoff when
radioactivity is detected in the molten salt conduits or when a
pressure change is detected in the molten salt conduits.
[0029] FIG. 4 shows another embodiment of a molten salt nuclear
reactor in accordance with the present disclosure.
[0030] FIG. 5 shows a top view of the embodiment of FIG. 4.
[0031] FIG. 6 shows yet another embodiment of a molten salt nuclear
reactor in accordance with the present disclosure.
[0032] FIG. 7 shows a further embodiment of a molten salt nuclear
reactor in accordance with the present disclosure.
[0033] FIG. 8 shows an additional embodiment of a molten salt
nuclear reactor in accordance with the present disclosure.
[0034] FIG. 9 shows a flowchart of a method according to certain
examples of the present disclosure.
[0035] FIG. 10 shows a nuclear power plant according to certain
examples of the present disclosure.
[0036] FIG. 11 shows another embodiment of a nuclear reactor in
accordance with the present disclosure.
[0037] FIG. 12 shows yet another embodiment of a nuclear reactor in
accordance with the present disclosure.
DETAILED DESCRIPTION
[0038] The present disclosure provides an integral Molten Salt
Reactor (IMSR). The IMSR of the present disclosure has a graphite
core that is permanently integrated with the vessel of the IMSR,
which means that the graphite core is in the vessel of IMSR for the
lifetime of the IMSR. As such, in the IMSR of the present
disclosure, the graphite core is not a replaceable graphite core
and remains within the IMSR for the operational lifetime of the
IMSR. The graphite core is fixedly secured within the vessel of the
IMSR. Advantageously, this eliminates the need for any apparatus
that would be required for replacing the graphite core at
pre-determined moments as per a pre-determined schedule. A further
advantage is that the IMSR does not require any access port to
allow access to the graphite core for replacement of the graphite
core. An additional advantage of the IMSR of the present disclosure
is that, after expiration of the design lifetime of the IMSR, the
IMSR serves as a storage container for any radioactive matter
within the IMSR. The components of the IMSR include the reactor
vessel itself and any graphite elements of the nuclear core. Other
components can include the primary heat exchangers which can be
installed, in the reactor vessel, during fabrication of the IMSR.
The IMSR is built to operate (produce electricity) for a design
lifetime, which takes into account the reactor's graphite core
expansion over time and the structural integrity of the graphite
core. That is, as mentioned above in the background section, the
graphite core will eventually expand beyond its original volume
under neutron flux. Operation of MSRs in the presence of such
expansion is not desirable as the graphite core can suffer breaks.
The IMSR of the present disclosure is simply shut down and replaced
after expiration of its design lifetime. Further components of the
IMSR can include piping such as coolant salt inlet conduits and
outlet conduits, and the pump shaft and impeller for moving
(pumping) the coolant salt (primary coolant fluid) when a pump is
employed.
[0039] In some embodiments of the present disclosure, an IMSR that
has been shut down can simply remain in its containment zone (hot
cell) that can act as a heat sink for the decay heat generated by
the shut down IMSR. The decay heat simply radiates out the IMSR
through the IMSR's vessel wall and into the containment zone and
ultimately to the outside environment. MSRs typically operate at
temperatures in the region of 700 degrees C., radiant heat is very
effective in removing decay heat. Further, to accelerate decay heat
removal, the IMSR of the present disclosure, a buffer salt can be
added in the containment zone to surround the IMSR; this allows
faster heat extraction from the IMSR to the containment zone. In
certain embodiments the IMSR can have a frozen plug of salt that
can be melted to allow the primary coolant drain to decay heat
removal tanks.
[0040] In some other embodiment, during operation of the IMSR and
after shut down of the IMSR, the IMSR can be a sealed unit that
simply retains produced fission gases within the IMSR sealed vessel
or, the fission gases can be release slowly to any suitable fission
gases treatment system.
[0041] In the present disclosure, elements can be said to be
operationally connected to each other when, for example,
information in one element can be communicated to another element
through a connection between the elements. The connection can be an
electrical connection. Further, elements can be said to be
operationally connected when state of one element can be controlled
by, or related to a state of another element.
[0042] Further, in the present disclosure, elements can be said to
be in fluid communication when fluid present at one element can
flow to the other element.
[0043] FIG. 1A shows the frontal view of an embodiment of an IMSR
90 of the present disclosure. 100 is the reactor vessel itself,
made of Hastelloy.RTM. N, a high nickel alloy, or any other
suitable material such the molybdenum alloy TZM
(titanium-zirconium-molybdenum alloy). The reactor vessel 100 can
be referred to as a sealed reactor vessel in the sense that any
graphite core within the reactor vessel 100 is sealed therein; that
is, it meant to remain within the reactor vessel 100, and not be
replaced during the operational lifetime of the IMSR. As the IMSR
100 of the present disclosure can have a short design life (e.g., 5
years), the walls of the reactor vessel 100 can be thinner than
required for MSRs that have a 30+ year design life and can be
allowed to operate in a much higher neutron fluence, or at a higher
operating temperature than such long lifetime MSRs. 102 shows the
core or core region which can be a simple mass of graphite defining
channels 115 for a molten salt fuel 108 to flow through. The
channels can also be referred to as through holes. The core 102 can
also be referred to as core region, a graphite moderator core, and
a graphite neutron moderator core. As the core 102 of the
embodiment of FIG. 1A does not need to be replaced, the
construction of the core 102 can be simplified in that it does
require any structural features that would allow and/or facilitate
its removal from the vessel 100 or its replacement. 104 shows a
reflector (neutron reflector) to reflect neutrons toward the core
102 and to shield the primary heat exchanger unit 106 from
excessive neutron flux. The reflector 104 can be optional. In the
absence of the reflector 104 any metallic structure, for example,
conduits and heat exchangers located in the IMSR above the core 102
would likely suffer neutron damage. The reflector 104 can be made
of stainless steel as it serves no structural purpose so
irradiation damage of the reflector 104 is of little concern. The
reflector 104 has channels 99 or piping defined therein to allow
the molten salt fuel 108 to flow from the primary heat exchanger
unit 106 through the channels 115 defined by the core 102. The
channels 115 can be varied in either diameter or lattice pitch in
different areas of the core 102 to create, for example, an
undermoderated region as well as an outer reflector zone in the
graphite, as would be understood by those trained in the field. In
the IMSR example of FIG. 1A, the flow of the molten salt fuel 108
in the vessel 100 is shown by the arrows 109.
[0044] The primary heat exchanger unit 106 has an opening 117 that
receives the fuel salt 109 provided by the drive shaft and impeller
unit 116, which is driven by a pump 118. The primary heat exchanger
unit 106 contains a series of heat exchangers. Such a heat
exchanger is shown at reference numeral 119. Each heat exchanger
119 is connected to an inlet conduit 114 and an outlet conduit 112
that propagate a coolant salt 113 (which can also be referred to as
a secondary coolant salt) from the outside of the vessel 100,
through the heat exchanger 119, to the outside of the vessel 100.
The coolant salt 113 flows through the inlet conduit 114, heat
exchanger 119, and outlet conduit 112 in the direction depicted by
arrows 111. The coolant salt 113 receives heat from heat exchanger
119, which receives the heat from the fuel salt 108 that flows on,
or circulates around, the heat exchanger 119. The secondary coolant
salt 113 is pumped by a pump or pumping system (not shown). For
clarity purposes, the heat exchanger 119 is shown as a straight
conduit connecting the inlet conduit 114 to the outlet conduit;
however, as would be understood by the skilled worker, the heat
exchanger 119 can be of any suitable shape and can include any
number of conduits connecting the inlet conduit 114 to the outlet
conduit 112. As an example, a heat exchanger can have a manifold
structure where coolant salt circulating in a main conduit is
divided into a plurality of conduits stemming from the main
conduit. Further, each heat exchanger can be individually shut down
upon occurrence of a heat exchanger fault and the nuclear reactor
can continue to operate with a reduced number of functioning heat
exchangers.
[0045] The heat exchanger unit 106, the heat exchangers 119 it
comprises, and the inlet conduits 114 and outlet conduits 112
connected to the heat exchangers 119 are all part of a heat
exchanger system that is used to transfer heat from the IMSR to a
system (an end use system) or apparatus such as, for example, a
steam generator. Such a heat exchanger system is shown elsewhere in
the disclosure, in relation to a nuclear power plant. The inlet
conduits 114 and the outlet conduits 112 are operationally
connected to a pump system--not shown--which is also part of the
heat exchanger system. That is, the pump system circulates the
coolant salt through the inlet conduits 114, the outlet conduits
112, and the heat exchangers 119. The inlet conduits 114 and the
outlet conduits 112 can be operationally connected to additional
heat exchangers that provide the heat of the coolant salt
circulating the heat exchangers 119, the inlet conduits 114 and the
outlet conduits 112 to another medium, such as, for example,
another fluid such as water.
[0046] In the example of FIG. 1A, the heat exchanger system is
partly comprised in the vessel 100 as the heat exchangers 119 and a
portion of inlet conduit 114 and the inlet conduit 112 are inside
the vessel 100. Further, the heat exchanger system is partly
outside the vessel 100 in that another portion of the inlet conduit
114 and the outlet conduit 112 are outside the vessel 100, as are
the aforementioned pump system and any additional heat exchanger.
That is to say, that the heat exchanger system has an inside
portion located inside the vessel 100, and an outside portion
located outside the vessel 100.
[0047] Also in the example of FIG. 1A, the molten fuel salt
circulates only in the vessel 100. That is, under normal operating
conditions, that is, conditions in which no break in equipment
occurs, the molten fuel salt 108 does not leave the vessel 100.
[0048] The IMSR 90 is positioned in a hot cell whose function is to
prevent radiation or radioactive elements, present or generated in
the IMSR 90, from traversing the hot cell walls. Such a hot cell
cell wall is partly shown at reference numeral 130. The outlet
conduit 112, and the inlet conduit 114, can pass through openings
in the hot cell wall 130 and can reach a secondary heat exchanger
(not depicted) giving heat to either a third loop of working fluid
or to the final working media such as steam or gas.
[0049] The level of molten fuel salt 108 within the reactor vessel
is depicted by reference numeral 122. Fission gasses will collect
above this liquid level 112 and may be retained in the vessel 100
or be allowed to transit, through an off gas line 120, to an off
gas sequestration area (not depicted). These off gasses can be
moved to the sequestration area by a helium entrainment system (not
depicted).
[0050] An example of the dimensions of the IMSR of FIG. 1A may be
3.5 meters in diameter, 7-9 meters in height, and may provide a
total power of 400 MW.sub.thermal (up to about 200
MW.sub.electrical). This power density would give a graphite
lifetime and thus design lifetime of the IMSR of somewhere between
5 and 10 years. These dimensions of the IMSR 90 make transport and
replacement of the IMSR 90 manageable and the power density allows
many years of usage of any graphite employed. The geometry of the
core 102 and vessel 100 can be cylindrical.
[0051] The core 102 can be fitted with, or connected to, one or
more stress monitors 902 that monitor the stress (shear stress,
normal stress, or both) that may develop in the core 102 over time,
as the core is subjected to neutrons. The stress monitors are
operationally connected to a control system 901 and, upon the
stress measured by the stress monitors 902 exceeding a
predetermined threshold value, the monitoring system can shut down
the IMSR 90. The one or more stress monitors (stress sensors,
strain sensors, stress detectors, stress gauges, strain gauges) can
include, for example, a ring surrounding the core with a strain
gauge connected (mounted) to the ring. Any overall expansion of the
graphite will create stress in the ring. The stress in the ring is
be detected by the strain gauge mounted on the ring. The one or
more stress monitors can also include a stress monitor mounted on
any other part that is secured to the core. For example, in
instances where the core is mounted to a mounting plate, a stress
monitor can be secured to the mounting plate. Stress in the core
will transfer to the mounting plate and will be sensed by the
stress sensor. The stress monitors can be, for example, electrical
in nature in that the resistance of the stress monitor will change
as a function of stress. The stress monitors may also be mechanical
or optical (e.g., optical fiber stress gauge).
[0052] In some embodiments, it is possible to determine the neutron
fluence on the core 102. That is, it is possible to determine the
number of neutrons per cm.sup.2 received by the core 102. It may
also be possible to monitor the fluence only for fast neutrons,
e.g., for neutrons having an energy above a particular energy level
(e.g., 50 KeV). One possible method of determining the neutron
fluence would be by inferring the neutron fluence by determining
(measuring) local power density which is directly related to both
fission power and fast neutron fluence. For example by placing
simple thermocouples separated by a short distance within a single
salt channel in the core, the temperature difference and flow rate
could be used to infer local power density. The IMSR can be shut
down automatically or manually when the total neutron fluence meets
a threshold criteria. For example, the IMSR can be shut down when
the neutron fluence approaches a pre-determined value beyond which
the core graphite 102 would likely deform or crack.
[0053] The IMSR 90 can be shutdown in any suitable manner. For
example, and with reference to FIG. 1B, upon occurrence of a
shutdown event such as excessive strain in the core 102 or
excessive neutron fluence on the core 102, the molten fuel salt 108
can be dumped in a dump tank 903 located below the vessel 90. Such
dump tanks can have any suitable geometry, provided the geometry in
question does not give rise to criticality. The dump tank 903 can
be connected to the vessel through any suitable valve mechanism
904. One such valve mechanism is freeze plug, which comprises a
portion of a conduit connecting the vessel 30 to the dump tank. The
portion of the conduit is filled with a material that is maintained
in the solid state by powered cooling (not shown). The material can
be a portion of the fuel salt itself. When the cooling stops, for
whatever reason such as controlled shutdown or a loss of external
cooling of the reactor, the material melts, opening the valve
mechanism 902, and the molten fuel salt 108 falls into the dump
tank 903.
[0054] Another example of a valve mechanism 904 is that of a
mechanical valve held in the open position by springs, and held in
the closed position by powered solenoids (not shown). As with power
of the powered cooling being remove or lost when power is cut or
lost in the solenoids, the solenoids will de-energize and the valve
will revert to its open position, under the force of the springs,
and the molten fuel salt will fall into the dump tank.
[0055] In the freeze plug example and the mechanical valve example,
the control system 901 would cut-off power to, respectively, the
cooling unit and the solenoids upon occurrence of a shutdown event
such as stress in the core 102, or excessive neutron fluence at the
core 102, or when external cooling is lost (failure/shutdown of the
heat exchanger system).
[0056] As another example, upon detection of a shutdown event, the
control system 901 can cause a control rod 902 to be lowered in the
vessel 90. The control rod 905 can be maintained out of the vessel
90 by a powered device 906 (e.g., a powered solenoid arrangement)
as long as there is power provided to the powered device. Upon
occurrence of a shutdown event or loss of external cooling of the
reactor, the control system 901 shuts off the power to the powered
device and the control rod lowers in the vessel 90.
[0057] FIG. 2 shows a top down view of the top of an example of an
IMSR of the present disclosure. FIG. 2 shows the pump motor 118,
and the off gas line 120. As well, FIG. 2 shows a series of four
inlet conduits 114 and four outlet conduits 114 passing from the
reactor vessel 100 through the primary hot cell wall 130. Four
separate pairs of lines (one pair of lines has one inlet conduit
114 and one outlet conduit 112) are depicted; however, any suitable
number of such pairs of lines (and associated heat exchanger 119)
is also within the scope of the present disclosure. Each pair of
lines is connected to a heat exchanger comprised in the heat
exchanger unit 106.
[0058] An advantage of keeping primary heat exchangers within the
IMSR and simply replacing the IMSR after its design lifetime, is
that techniques for heat exchanger repair, removal, and/or
replacement need not be developed. However plans must be made for
potential failure and leakage between the primary fuel salt and
secondary coolant. By compartmentalising the primary heat exchanger
unit 106 into multiple independent heat exchangers 119, any failure
of the heat exchangers 119 and/or leakage of molten fuel salt 108
into the coolant 113 can be effectively managed.
[0059] FIG. 3 shows an embodiment of a disconnect arrangement to
cut off the flow of the secondary coolant 113 though the inlet
conduits 114 and outlet conduits 112 in the direction given by
arrows 111. For clarity purposes, only one pair of lines (one inlet
conduit 114 and one outlet conduit 112) is shown in FIG. 3. In the
example of FIG. 3, a radioactivity detector 300, for example, a
Geiger counter is placed next to an outlet line 112 and can detect
any leak of radioactive primary fuel salt into the outlet line 112.
When radioactivity beyond a pre-determined level is detected by the
radioactivity detector 300, a controller 301, connected to the
radiation detector 301, controls shutoff mechanisms 304 that are
connected to the outlet conduit 112 and the inlet conduit 114, to
shut the outlet conduit 112 and its corresponding inlet conduit
114. The shutoff mechanisms are to isolate the individual heat
exchanger 119 (not shown in FIG. 2) connected to the now shut inlet
conduit 114 and outlet conduit 112. The shutoff mechanisms 304 can
also be to sever the physical connection along the inlet conduit
114 and the outlet conduit 112. The shutoff mechanisms can include
any suitable type of shutoff valves and any suitable type of
crimping devices, the latter to crimp shut the inlet conduit 114
and the inlet conduit 112. The shutoff mechanisms 304 can also
include a refrigerating unit that can cool and freeze the coolant
salt circulating in a compromised inlet conduit or a compromised
outlet conduit. Such freezing would occur in a segment of the
compromised conduit (inlet or outlet) and stop the flow of coolant
salt. In some embodiments, where the inlet and/or outlet conduits
are substantial in diameter and hence difficult to freeze, the
conduits can be mechanically stretched to reduce their diameter and
the sections of the conduits having the reduced diameter can be
frozen.
[0060] Further, if a leak of secondary coolant fluid 113 into the
molten fuel salt 108 occurs, it can be detected by measuring a drop
in pressure, using one or more pressure detectors 303 mounted in or
otherwise operationally connected to the inlet conduit 114, the
outlet conduit 112 or both. The one or more pressure detectors are
operationally connected to the controller 301, which can shut off
the shutoff mechanisms 304 upon determining that a drop in pressure
(or any abnormal change in pressure) has occurred in the coolant
salt 113 circulating in the inlet conduit 114, outlet conduit 112,
or both. Furthermore, when a leak of secondary coolant fluid 113
into the molten fuel salt 108 occurs, it can be detected by
monitoring (e.g., periodically monitoring) the level of molten salt
in the reactor vessel. If the level of molten salt rises, then it
can be attributed to a leak of secondary coolant salt.
[0061] In some embodiments, each pair or group of pairs of inlet
conduit and outlet conduit can be connected to a distinct coolant
pump. When a fault is detected in one of the pairs, the pump to
which the pair is associated can be shut down and the conduit in
question can be crimped, frozen or otherwise disabled by a shutoff
mechanism. Provided that all the coolant pumps are not shutdown,
the nuclear reactor can still function.
[0062] By choosing compatible primary carrier salts for the molten
fuel salt 108 and the secondary coolant salt 113, mixing of these
fluids can be tolerated. For example, if the primary carrier salt
is LiF--BeF2 and/or NaF--BeF2, then a secondary coolant salt of
LiF--BeF2 and/or NaF--BeF2 would be compatible with the primary
carrier salt in cases of limited mixing, i.e. in cases where the
volume of coolant salt 113 leaked in into the molten fuel salt 108
is tolerable in terms of its effects on neutron production and
absorption. By having many, perhaps 4 but even up to 10 or more
pairs of inlet conduits/outlet conduits (and corresponding heat
exchangers 119), the loss of one or more individual heat exchangers
may do little to the overall ability to transfer heat from the
primary heat exchanger unit 106 to the coolant salt 113 as the
other remaining pairs of inlet conduits/outlet conduits can simply
take the added heat exchange load or the IMSR can lower its power
rating slightly. Heat exchangers are unlike many other systems in
that there is very little economy of scale such that 10 smaller
pairs of inlet/outlets or tube bundles will not have a combined
cost much more than one large unit.
[0063] FIG. 4 shows another embodiment of an IMSR 92 in accordance
with the present disclosure. As in the IMSR 90 of FIG. 1A, the IMSR
92 of FIG. 4 comprises a vessel 100, a reflector 104 and a core
102. Additionally, the IMSR 92 comprises a control rod 400 (which
can be optional) and a series of heat exchanger units 106. Each
heat exchanger unit has a drive shaft and impeller unit 116 to pump
molten fuel salt 108 through the heat exchanger units 106. For
clarity purposes, pump motors that drive the shaft and impeller
units 116 are not shown. Also for clarity purposes, inlet conduits
and outlets conduits propagating a coolant salt through the heat
exchanger units 106 are not shown.
[0064] The molten salt fuel 108 that is pumped through the heat
exchanger units 106 is directed downwards, towards the periphery of
the core 102 by a baffle structure 402. The molten fuel salt flows
towards the bottom of the vessel 100 and then upwards through the
channels 115 of the core 102. Although two channels 115 are shown
in FIG. 4, any suitable number of channels 115 is within the scope
of the present disclosure.
[0065] FIG. 5 shows a top, cross-sectional view of the MSR 92 shown
at FIG. 4. The top view of FIG. 5 shows 8 heat exchanger units 106,
each having an inlet conduit 114, an outlet conduit 112, and a pump
shaft and impeller unit 116. Also shown is the control rod 400.
[0066] FIG. 6 shows a side perspective view of the IMSR of FIG. 4.
The IMSR 92 comprises six heat exchanger units 106, each having an
inlet conduit 114, outlet conduit 112, and shaft and impeller unit
116. The heat exchanger units 106 are positioned above the core 102
and about a longitudinal axis of the vessel, the longitudinal axis
being parallel to the control rod 400. The direction of flow of the
molten fuel salt 108 is indicated by arrow 109. After exiting the
individual heat exchangers 106, the molten fuel 108 flows obliquely
down, guided by the baffle structure 402 and, optionally, by
partitions 404 that separate the outputs of the individual heat
exchanger units.
[0067] The flow of the molten fuel salt 108 through the core 102
may be in different directions in different embodiments, for
example upwards as shown in the embodiment of FIG. 4 or downwards
as shown in the embodiment of FIG. 1A. There are advantages and
disadvantages to both upwards and downwards flow directions. An
upward flow through the core as shown in FIG. 4 has the advantage
of being in the same direction as natural circulation but can make
the use of pumps (the pumps pumping the coolant salt through the
heat exchanger units) slightly more difficult to direct the flow
through the primary heat exchangers.
[0068] In some embodiments of the present disclosure, the pumps and
the shaft and impeller units can be omitted and the MSR can instead
use natural circulation to circulate the molten fuel salt 108. As
such, the pumps and the shaft and impeller units can be optional in
embodiments where natural circulation suffices to circulate the
molten salt fuel 108. FIG. 7 shows an embodiment where natural
circulation of the molten fuel salt 108 is used. The MSR 94 of FIG.
7 is similar to the MSR 92 of FIG. 6 with the exception that no
pumps or shaft an impeller units are required. Rather, the molten
fuel salt 108 present in the channels 115 heats up through nuclear
fission reaction and flows upwards towards the top region of the
vessel 100. Once outside the channels 115, the molten salt cools
down and begins to flow downwards, through the heat exchangers 105,
and towards the bottom of the vessel 100 where the cooled molten
fuel salt re-enters the channels to be heated up.
[0069] FIG. 8 shows another embodiment of an IMSR in accordance
with the present disclosure. The IMSR 96 of FIG. 8 has a vessel 100
in which is positioned a graphite moderator core 102, which can
have one or more channels 115 defined therein. The vessel 100 is
connected to a heat exchanger unit 106 that is located outside the
vessel 100. The heat exchanger unit 106 contains a plurality of
heat exchangers (not shown); each heat exchanger includes an inlet
conduit 114 and an outlet conduit 112 that circulate coolant salt
though the heat exchanger. Each inlet conduit 114 and outlet
conduit 112 is operationally connected to a coolant salt pump
system (not shown). The inlet conduit 114 and the outlet conduit
112 are shown traversing a hot cell wall 130. The vessel 100 is
connected to the heat exchanger unit 106 through conduits 700 and
702. A pump 704 circulates a molten fuel salt 706 through the
vessel 100, the channels 115, and the heat exchanger 106. The same
configuration of radioactivity detector, pressure detectors 303,
shut-off mechanisms, and controller shown at FIG. 3, can also be
applied to the embodiment of FIG. 8. The core 102 can be fitted
with one or more stress monitors 902 that monitor the stress (shear
stress, normal stress, or both) that may develop in the core 102
over time, as the core is subjected to neutrons. The stress
monitors are operationally connected a monitoring system (not shown
and, upon the stress measured by the stress monitors 902 exceeding
a predetermined threshold value, the monitoring system can shut
down the IMSR 96.
[0070] Upon the graphite moderator core 102 reaching its
operational lifetime, the conduits 700 and 702 can be severed to
physically disconnect the vessel 100 from the remainder of the
IMSR. After sealing the cut-off portion of the conduits 700 and 702
attached to the vessel 100, the vessel 100 can be disposed in a
containment facility and a new vessel with a new graphite moderator
core can be attached to the conduits 700 and 702.
[0071] The IMSR embodiments shown at FIGS. 1-8 were described has
having a molten fuel salt (108 or 706) circulating therein.
However, modifications to the embodiments of FIGS. 1-8 would allow
the IMSRs shown therein to operate on a solid nuclear fuel
comprised within the core 102 as opposed to being comprised in the
molten fuel salt. For example, in the embodiment of FIG. 1A, the
molten fuel salt can be replaced by a fuel-free (nuclear fuel-free)
molten salt and the core 102 can comprise solid nuclear fuel such
as TRISO fuels. Further, as no fission gasses are released in such
solid fuel IMSRs, there would be no need for the off gas line 120.
As previously described however, there are similar advantages to
the invention of integrating a sealed solid fuel core into the
replaceable IMSR unit.
[0072] FIG. 9 shows a block diagram of an embodiment of a nuclear
power plant 2000 that includes an MSR 2002 such as, any one of IMSR
90, 92, 94, and 96 described above in relation to FIGS. 1, 4, 6, 7,
and 8. The MSR 2002 generates heat and provides the generated heat
to a heat exchanger system 2004. The heat exchanger system 2004 can
include the heat exchanger unit 106 disposed in the vessel 100,
which also includes a graphite moderator core 102 and is discussed
above in relation FIGS. 1, 4, 6, and 7. With respect to MSR 96
shown at FIG. 8, the heat exchanger system 2004 can include the
heat exchanger unit 106, which is located outside the vessel 100
that includes the graphite moderator core 102. Additionally, the
heat exchanger system 2004 of FIG. 9 can include additional heat
exchangers that receive the heat from the above noted heat
exchanger units 106. The nuclear power plant 2000 of FIG. 9
includes an end-use system 2006 that receives heat from the heat
exchanger system 2004 and uses that heat to do work. For example,
the end-use system 2006 can include a heat exchanger apparatus that
transport the heat received from the heat exchanger system 2004 to
an industrial apparatus that uses that heat. An example of such an
industrial apparatus includes a cement kiln. In other embodiments,
the end-use system 2006 can include a steam generator that uses the
heat received from the heat exchanger system 2004 to produce steam
that powers a turbine system, which can be used to power an
electrical generator. In further embodiments, the end-use system
2006 can include a steam generator that uses the heat received from
the heat exchanger system 2004 to produce steam that is used for
bitumen extraction from bituminous sands (e.g., steam assisted
gravity drainage).
[0073] FIG. 10 shows a flowchart of a method according to certain
examples of the present disclosure. The method shown at FIG. 10 is
a method of operating a nuclear power plant. The nuclear power
plant comprises a nuclear reactor (e.g., an MSR) that generates
heat (thermal energy) and a heat exchanger system. The nuclear
reactor comprises a vessel, a graphite moderator core positioned in
the vessel, and a molten salt circulating at least in the vessel.
In embodiments where the nuclear reactor is an MSR, the molten salt
is a molten fuel salt. The nuclear reactor heats the molten salt
and the heat exchanger system receives the heat from the molten
salt.
[0074] The method of FIG. 10 includes, at action 1000, operating
the nuclear reactor. At action 1002, the MSR is shut down upon
occurrence of a shutdown event. Shutdown events can include, for
example, a detection of strain in the graphite moderator core the
neutron fluence on the graphite moderator exceeding a maximum
fluence level, and an operation duration of the nuclear reactor
exceeding a pre-determined operation duration. The pre-determined
duration of operation is determined in relation to maintaining the
structural integrity of the graphite moderator core positioned in
the vessel of the MSR and in relation to the operation conditions
under which the MSR operates. For a given graphite moderator core,
when the pre-determined operation conditions are such that the
graphite moderator core is subjected to low peak power densities
and low average power densities, the pre-determined duration of
operation will be longer than when the pre-determined operation
conditions are such that the graphite moderator core is subjected
to high peak power densities and high average power densities. An
MSR having a peak power density of 20 MW.sub.thermal/m.sup.3 would
result in the pre-determined duration of operation being about 11.5
years when running at full capacity, and about 15 years when
running at 75% capacity. It is envisaged that the operational time
(duration) of a practical IMSR will be less than 15 years and thus,
will have a peak power density higher than 20
MW.sub.thermal/m.sup.3.
[0075] At action 1004, all operational connections between the
inside portion of the heat exchanger system and the outside portion
of the heat exchanger system are severed. This results in a
severed, shut-down nuclear reactor. That is, any type of conduit
connected to the nuclear and used to transfer heat from the nuclear
reactor to any part of the heat exchanger system located outside
the vessel is severed. Further, electrical connections for pump
motors and monitoring instrumentation, small conduits for makeup
fuel salt addition, salt sampling, off gas removal and a dip line
for the removal of the fuel salt can also be severed when, for
example, the severed shutdown nuclear reactor is to be moved or
sequestered
[0076] At action 1006, a replacement nuclear reactor can be
obtained and, at action 1008, the inner heat exchanger system
portion of the replacement nuclear reactor is connected to the
outside portion of the heat exchanger system. If applicable, any
other electrical connections for pump motors and monitoring
instrumentation, small conduits for makeup fuel salt addition, salt
sampling, off gas removal and a dip line for the removal of the
fuel salt of the replacement nuclear reactor can be made.
[0077] At action 1001, if fault in a heat exchanger is detected,
the flow of coolant salt in the faulty heat exchanger can be
stopped. At action 1005, the severed, shutdown nuclear reactor can
be sequestered.
[0078] To shut down the nuclear reactor, a control rod (shutdown
rod) can be used or, in embodiments where the nuclear reactor is an
MSR, by draining the molten fuel salt to an external storage such
as a dump tank. The coolant lines can then be sealed and/or crimped
and disconnected along with any other lines such as off gas lines.
Examples of coolant lines are shown in FIG. 1 as inlet conduit 114
and outlet conduit 112. After disconnecting these lines the spent
nuclear reactor, i.e., the reactor vessel and all remaining conduit
segments attached thereto, can be removed, for example, by using an
overhead crane. Such operations might be done after a period of in
situ cool down for radiation levels to diminish. In such a mode,
likely the next unit (i.e., the replacement nuclear reactor) can be
installed adjacent the spent IMSR such that, long term, while one
unit operates, the other is cooling down and then replaced before
the operating unit is finished its cycle. Using an overhead crane
for removal may involve some mechanism to breach the primary hot
cell.
[0079] The pump motor (see reference numeral in FIG. 1), when
present, can be recycled, for example by, cutting it from the shaft
of the impeller to which the pump motor is connected. The rest of
the spent nuclear reactor can be transferred off site or to another
area of the nuclear power plant, perhaps even within the primary
hot cell. As an option, the unit might also be used for the short,
medium or even long term storage of the primary fuel salt itself,
perhaps after some or all actinides are removed for recycle or
alternate storage. Thus the spent nuclear reactor may act as a
storage and/or disposal canister for the internal graphite, primary
heat exchangers and even the salt itself. At some point a decision
on long term sequestration would have to be made but potentially
the entire unit could be lowered into an underground location such
as deep borehole made on site or transported to a salt cavern for
safe long term sequestration.
[0080] Some comment on the overall economic viability is perhaps of
use as it goes against the often imposed logic of attempting to get
the longest service life as possible from all components. The
advantages seem to greatly outweigh any economic penalty of
decreased capital amortization time. First, there may be little
change in the overall need of graphite over the lifetime of the
nuclear plant itself as would be understood by those trained in the
field. Second, the components now having a shorter design life such
as the reactor vessel and/or primary heat exchangers typically make
up only a small fraction of the nuclear plant costs. In studies by
Oak Ridge National Laboratories, such as in ORNL 4145 the cost of
the reactor vessel and primary heat exchangers were only around 10%
of the plant cost. The ability to lower the cost of these items by
the great simplifications allowed by having a sealed replaceable
unit would seem to more than make up for the lowered amortization
time. When the decreased research and development costs are
factored in, the advantage of this disclosed design seem clear.
[0081] FIG. 11 shows a top, cross-sectional view of a further
embodiment of a nuclear reactor 1100 of the present disclosure. The
nuclear reactor 1100 has a nuclear reactor vessel, which has a
nuclear reactor vessel wall 1104 and, the nuclear reactor vessel
1102 is contained in a containment vessel 1106, which has a
containment vessel wall 1108. Between the nuclear reactor vessel
wall 1104 and the containment vessel wall 1108 is a buffer salt
1110. The nuclear reactor wall 1104 is made of a thermally
conductive material, for example, a nickel-base alloy such as
Hastelloy.RTM. N. The buffer salt 1110 is in thermal contact with
the nuclear reactor wall 1104.
[0082] Upon loss of electrical power to the heat exchanger system,
the pumps pumping the coolant salt through the heat exchangers
located inside the vessel will stop functioning. However, some of
decay heat will continue to be transferred out the reactor vessel
through natural circulation: that is, the coolant salt in the
reactor vessel will heat up and circulate through the secondary
heat exchangers (secondary heat exchanger loops) system by
convection. As such, provided the heat exchanger system remains
able to shed some of the heat received by nuclear reactor, severe
consequences, such as damaging the metallic structure of the
nuclear reactor vessel, can be avoided.
[0083] However, upon a catastrophic event, for example an
earthquake, where the heat exchanger system becomes thoroughly
defective, i.e., is no longer able to transfer any significant heat
from the nuclear reactor 1102, the nuclear reactor 1102 can no
longer transfer the decay heat generated therein and failure to
properly manage the decay heat can lead to severe consequences.
[0084] In accordance with the present disclosure, the decay heat
can be safely managed by selecting a buffer salt 1110 that acts as
a phase transition heat sink. When used in MSRs, the buffer salt
provides an alternative to the freeze plug and dump tank approach
often used in MSRs. The virtue of the embodiment of FIG. 11 is the
ability to passively dissipate the decay heat that is produced by
nuclear reactors after the loss of external cooling (i.e., when the
heat exchanger system can no longer transfer any significant heat
from the nuclear reactor). The embodiment of FIG. 11 enables the
dissipation of the decay-heat surge even when there is loss of
external cooling, thereby avoiding severe consequences.
[0085] As an example, the nuclear reactor 1100 can be considered to
be an MSR that runs at about 650.degree. C. and produces thermal
energy at a rate of 80 MW.sub.th (full power value) and the nuclear
reactor vessel wall 1104 is at 650.degree. C. Upon shutdown, the
decay heat generated by the nuclear reactor will be, averaged over
the first two days, about 0.5% of the full power value and the
temperature of the nuclear reactor vessel wall 1104 will
increase.
[0086] When the buffer salt 1110 is 53% NaF-47% AlF.sub.3 (density
of 2.4 t/m.sup.3 with 400 kJ/kg latent heat, melting point of
695.degree. C.) and is 1 meter thick, the total mass of the buffer
salt is about 177 tons and provides a latent heat of melting of
7.1.times.10.sup.10 joules. In this example, the buffer salt 1110
provides approximately 2 days of initial decay heat absorption even
with an adiabatic assumption of no other heat loss. That is, it
will take about two days for the buffer salt 1110 to melt, i.e.,
about two days for the temperature of the nuclear reactor vessel
wall 1104 and of the buffer salt 1110 to reach the buffer salt's
melting point of 695.degree. C.
[0087] After the buffer salt has melted it remains in the
containment vessel 1106, surrounding the nuclear reactor 1102, the
decay heat is no longer absorbed by the buffer salt and needs to me
managed otherwise. Several options of managing the decay heat are
available. For example, the containment vessel can be surrounded by
water (a water jacket) 1112 that will be boiled off by the decay
heat. In the present example the water 1112 will boil off at a rate
of about 8 liters/minute (this boil-off rate will decrease with
time as less and less decay heat is generated). The boiled off
water can be replenished by a water reservoir (not shown). A modest
reservoir can supply water for many months, especially in view of
the unrealistic adiabatic assumption; clearly, radiant and
conductive heat will be dissipated into the building housing the
nuclear and in the environment surrounding the water jacket. As
such, the realistic water boil-off rate will be less that 8
liters/minute. The water jacket can be in the form of coiled piping
surrounding the containment vessel and in thermal contact with the
containment vessel wall 1108. The coiled piping is connected to the
water reservoir. In other embodiments, an air jacket can be used.
The air jacket can be in the form of coiled piping surrounding the
containment vessel and in thermal contact with the containment
vessel wall 1108. As will be understood by the skilled worker, in
some embodiments, providing cooling to the containment vessel may
cause a relatively thin layer of the buffer salt adjoining the
outside wall of the containment vessel to remain in the solid state
when the temperature at the wall in question is at, or below, the
freezing point of the buffer salt. Such embodiments are within the
scope of the present disclosure.
[0088] The buffer salt 1110 can be selected to be a thermal
insulator when in the solid state and a thermal conductor when in
the liquid (molten buffer salt) state. Specifically, the solid
state thermal conductivity of the selected buffer salt is lower
than the heat transfer capability of the liquid state buffer salt.
That is, convective heat transfer in the liquid state is
significantly higher than conductive heat transfer in the solid
state. 53% NaF-47% AlF.sub.3 is such a buffer salt. Having the
buffer salt 1110 acting as a thermal insulator during operation of
the nuclear reactor reduced loss of heat generated by the nuclear
reactions taking place in the nuclear reactor vessel 1102.
[0089] FIG. 12 shows a top, cross-sectional view of a further
embodiment of a nuclear reactor 1114 of the present disclosure. As
in the nuclear reactor 1100 of FIG. 11, the nuclear reactor 1114
has a nuclear reactor vessel 1102, which has a nuclear reactor
vessel wall 1104 and, the nuclear reactor vessel 1102 is contained
in a containment vessel 1106, which has a containment vessel wall
1108, which can be referred to as an outer wall or as a containment
vessel outer wall. Additionally, the containment vessel has an
inner wall 116 (shown with dashed line) that is in thermal contact
with the nuclear reactor vessel wall 1104. Between the inner wall
1116 and the containment vessel wall 1108 is the buffer salt 1110.
The inner wall 1116 is thermally conductive and, as such, the
buffer salt 1110 is in thermal contact with the nuclear reactor
wall 1104. Advantageously, the nuclear reactor 1114 allows for
removal of the nuclear reactor vessel 1102 from the containment
vessel 1106 without having to remove the buffer salt 1110. Also, a
replacement nuclear reactor can be inserted in the containment
vessel 1106.
[0090] Even though the above examples use 53% NaF-47% AlF.sub.3 as
a buffer salt, any other suitable buffer salt can be used. That is,
salts that have a melting point above the operating temperature of
the nuclear reactor and that can act as a thermal insulator in the
solid state and as a thermal conductor (by convection) in the
liquid state can be used. Other examples of salts that can be used
as buffer salts include: other fluoride salts such as 66% NaF-34%
ZrF4 (melting point of 640.degree. C.) and 26% KF-74% Zr4 (melting
point of 700.degree. C.); bromide salts such as NaBr (melting point
of 747.degree. C., latent heat of melting: 250 KJ/Kg) and KBr
(melting point of 734.degree. C.; and other salts such as MgCl
(melting point of 714.degree. C., latent heat of melting: 360
kJ/Kg).
[0091] Even though the nuclear reactors of FIGS. 11 and 12 are
shown with buffer salts, other embodiments may use a buffer
material other than a buffer salt. For example, the buffer salt
1110 of FIGS. 11 and 12 can be replaced by pure aluminum (melting
point of 660.degree. C., latent heat of melting: 397 kJ/Kg). In
this case, to avoid having excessive heat transfer between the
nuclear vessel and the containment vessel during normal operation
of the nuclear reactor, the aluminum can be in the form of balls,
which allows for only some thermal contact between neighbouring
balls and the nuclear reactor vessel wall and the containment
vessel.
[0092] As with other nuclear reactors described herein, the nuclear
reactors shown at FIGS. 11 and 12 can also be disconnected,
removed, and replaced as a unit, with or without the containment
vessel.
[0093] In the preceding description, for purposes of explanation,
numerous details are set forth in order to provide a thorough
understanding of the embodiments. However it will be apparent to
one skilled in the art that these specific details are not
required.
[0094] The above described embodiments are intended to be examples
only. Alterations, modifications and variations can be effected to
the particular embodiments by those skilled in the art without
departing from the scope, to be defined solely in the accompanying
claims.
* * * * *